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< Page ,Total 45 >
NECP-Atlas: A new nuclear data processing code EI Scopus SCIE
期刊论文 | 2019 , 123 , 153-161 | Annals of Nuclear Energy
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Abstract :

A new nuclear data processing code called NECP-Atlas is under development at Xi'an Jiaotong University in China. The motivation for the development is to establish a platform to carry out deeper researches on nuclear data processing methods to satisfy the demands on accurate cross sections in the fields of high-fidelity transport simulation and advanced reactor design. The present paper describes the methods used in the current version of NECP-Atlas and demonstrates the performance and accuracy of the code on a variety of benchmarks. At the present time, NECP-Atlas can process ENDF/B-VII.1, ENDF/B-VII.0, CENDL-3.1, JEFF-3.2 and JENDL-4.0 evaluations, and generate WIMS-D and ACE format libraries. The accuracy of the code is comparable with NJOY2016. NECP-Atlas is competent to provide cross sections for deterministic and Monte Carlo transport calculations. © 2018 Elsevier Ltd

Keyword :

Advanced reactor design Cross section Multi band NECP-Atlas Resonance elastic scattering Transport calculation Transport simulation Xi'an Jiaotong University

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GB/T 7714 Zu, Tiejun , Xu, Jialong , Tang, Yongqiang et al. NECP-Atlas: A new nuclear data processing code [J]. | Annals of Nuclear Energy , 2019 , 123 : 153-161 .
MLA Zu, Tiejun et al. "NECP-Atlas: A new nuclear data processing code" . | Annals of Nuclear Energy 123 (2019) : 153-161 .
APA Zu, Tiejun , Xu, Jialong , Tang, Yongqiang , Bi, Huchao , Zhao, Fei , Cao, Liangzhi et al. NECP-Atlas: A new nuclear data processing code . | Annals of Nuclear Energy , 2019 , 123 , 153-161 .
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Heterogeneous discontinuity factor treatment in Variational Nodal Method EI SCIE
期刊论文 | 2019 , 127 , 341-350 | Annals of Nuclear Energy
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Abstract :

To handle the control rod cusping effect in Pressurized Water Reactor (PWR) core calculation, the heterogeneous Variational Nodal Method (VNM) employed by the fuel management calculation code system NECP-Bamboo has been enhanced to treat heterogeneous discontinuity factor (DF) appearing on nodal interface. To solve the neutron-diffusion equations with heterogeneous DF, firstly, the functional in VNM is modified to contain the discontinuity of neutron flux in the surface integral term. Secondly, other than volumetric flux and surface partial currents, cross sections and surface DF are also expanded into the sum of orthogonal piece-wise polynomials to construct the nodal response matrixes. Four test problems in this paper including the CISE, Henry-Worley, HAFAS benchmark problems and the BEAVRS problems were employed to verify the method in treating heterogeneous DF. It has been demonstrated that the control rod cusping effect can be fully eliminated by the heterogeneous VNM with heterogeneous DF in terms of control rod differential worth and three-dimensional pin-power profile. © 2018 Elsevier Ltd

Keyword :

Bench-mark problems Core calculations Differential worth Discontinuity factor Neutron diffusion equations Partial currents Surface integrals Variational nodal method

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GB/T 7714 Li, Yunzhao , Liang, Boning , Wu, Hongchun et al. Heterogeneous discontinuity factor treatment in Variational Nodal Method [J]. | Annals of Nuclear Energy , 2019 , 127 : 341-350 .
MLA Li, Yunzhao et al. "Heterogeneous discontinuity factor treatment in Variational Nodal Method" . | Annals of Nuclear Energy 127 (2019) : 341-350 .
APA Li, Yunzhao , Liang, Boning , Wu, Hongchun , Li, Zhipeng , Yang, Jiewei . Heterogeneous discontinuity factor treatment in Variational Nodal Method . | Annals of Nuclear Energy , 2019 , 127 , 341-350 .
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The material-region-based 2D/1D transport method EI
期刊论文 | 2019 , 128 , 1-11 | Annals of Nuclear Energy
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The 2D/1D transport method is the dominant method for high-fidelity direct whole-core transport calculations, which attracts a lot of attention in recent decades. In the 2D/1D method, some sources of deviation are introduced, including spatial and angular approximation of leakage term and cross-section homogenization for 1D axial calculation. These approximations are analyzed and a material-region-based 2D/1D transport method with anisotropic leakage term, which avoids cross-section homogenization, as well as reduces spatially flat leakage approximation, are developed to improve accuracy at the expense of acceptable memory and efficiency loss. Finally, a BWR assembly case, the C5G7 benchmark and a rod-cluster assembly case are tested to verify the accuracy and performance of the material-region-based 2D/1D method. © 2018 Elsevier Ltd

Keyword :

Cluster assembly Efficiency loss High-fidelity Leakage terms Material region NECP-X Transport method Whole-core transport

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GB/T 7714 Liu, Zhouyu , Zhao, Chen , Cao, Lu et al. The material-region-based 2D/1D transport method [J]. | Annals of Nuclear Energy , 2019 , 128 : 1-11 .
MLA Liu, Zhouyu et al. "The material-region-based 2D/1D transport method" . | Annals of Nuclear Energy 128 (2019) : 1-11 .
APA Liu, Zhouyu , Zhao, Chen , Cao, Lu , Wu, Hongchun , Cao, Liangzhi . The material-region-based 2D/1D transport method . | Annals of Nuclear Energy , 2019 , 128 , 1-11 .
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An improved fitting method for subgroup parameters based on the heterogeneous cells EI SCIE
期刊论文 | 2019 , 56 (2) , 179-192 | Journal of Nuclear Science and Technology
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An accurate subgroup parameters fitting method, where background cross sections obtained based on heterogeneous cells are used to fit the subgroup level and subgroup weight, is proposed in this paper. Due to the dependence of background cross section on the subgroup level, the calculation of the subgroup parameters is a nonlinear problem, which causes the iteration between fitting subgroup parameters and updating background cross sections. The cubic spline interpolation method is used to update the background cross sections to avoid frequently solving fixed source equations. In the fitting process, the negative subgroup parameters are often obtained, and the accuracy of the subgroup parameters is very sensitive to the iterative initial values of subgroup levels. To avoid these problems, additional constraints ensuring positive subgroup parameters and guaranteeing numerical stability are added to the optimization function. Penalty function method is used to convert the optimization problem with constraints into the one without constraints, making the problem easy to be solved. The proposed method is tested against the problems of pin cell, pressurized water reactor assemblies and plate-type assembly. The numerical results show that the self-shielded cross sections calculated by the proposed method agree well with those by Monte Carlo code. © 2018, © 2018 Atomic Energy Society of Japan. All rights reserved.

Keyword :

Cubic spline interpolation method Fitting method Nonlinear problems Optimization function Optimization problems Penalty function methods Subgroup method subgroup parameters

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GB/T 7714 Zu, Tiejun , Xia, Fan , Wu, Hongchun . An improved fitting method for subgroup parameters based on the heterogeneous cells [J]. | Journal of Nuclear Science and Technology , 2019 , 56 (2) : 179-192 .
MLA Zu, Tiejun et al. "An improved fitting method for subgroup parameters based on the heterogeneous cells" . | Journal of Nuclear Science and Technology 56 . 2 (2019) : 179-192 .
APA Zu, Tiejun , Xia, Fan , Wu, Hongchun . An improved fitting method for subgroup parameters based on the heterogeneous cells . | Journal of Nuclear Science and Technology , 2019 , 56 (2) , 179-192 .
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Validation of SARAX for the China Fast Reactor with the extrapolated experimental data EI SCIE
期刊论文 | 2019 , 127 , 188-195 | Annals of Nuclear Energy
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Abstract :

The validation works have been implemented to a newly-developed code SARAX for China Fast Reactor (CFR) in this paper. Different with the conventional way to validate the code using the dedicated experimental data, a theoretical approach was proposed and implemented by using the existing similar experimental data. This theoretical approach is based on the technology of sensitivity/uncertainty analysis, similarity analysis and nuclear-data adjustment. In our previous works, detailed introduction towards sensitivity/uncertainty analysis and similarity analysis have been implemented to distinguish the existing experiments which are 'similar’ to the CFR's criticality characteristics. This paper focus on the method to extrapolate the 'similar’ experimental data to predict the best-estimate measurements of CFR with application of nuclear-data adjustment, providing reference values to validate the SARAX code. From the numerical results, it can be observed that through nuclear-data adjustment, the simulation results of the existing experiments can agree well with corresponding measurements, with the bias reduced notably to be within 25 pcm. Moreover, the nuclear-data adjustment can also improve the nuclear-data uncertainties notably, with the relative uncertainties of the reactor keff due to nuclear data having been reduced from the value exceeding 1.0% to the values about 0.15%. © 2018

Keyword :

Code validation Nuclear data Numerical results Reference values Relative uncertainty Sensitivity/uncertainty analysis Similarity analysis Theoretical approach

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GB/T 7714 Wan, Chenghui , Qiao, Liang , Zheng, Youqi et al. Validation of SARAX for the China Fast Reactor with the extrapolated experimental data [J]. | Annals of Nuclear Energy , 2019 , 127 : 188-195 .
MLA Wan, Chenghui et al. "Validation of SARAX for the China Fast Reactor with the extrapolated experimental data" . | Annals of Nuclear Energy 127 (2019) : 188-195 .
APA Wan, Chenghui , Qiao, Liang , Zheng, Youqi , Cao, Liangzhi , Wu, Hongchun . Validation of SARAX for the China Fast Reactor with the extrapolated experimental data . | Annals of Nuclear Energy , 2019 , 127 , 188-195 .
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Improvement of Multi-group Cross Section Data Library of Bamboo-Lattice Code EI Scopus CSCD PKU
期刊论文 | 2018 , 52 (8) , 1367-1373 | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology
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Abstract :

Based on the ENDF/B-Ⅶ.0 nuclear evaluation data library, an improved multi-group cross section data library NECL2.0 was produced for Bamboo-Lattice code by using the nuclear data processing code NJOY and LATTICE_PRE. The results based on benchmark and numerical analysis show that the calculation results of infinite multiplication factor kinf, fission rate distribution and few-group homogenized cross section by using NECL2.0 library agree well with the corresponding reference values. Considering the resonance of Ag-In-Cd can increase accuracy of kinf by 1 000 pcm. Compared with reference values, the maximum relative deviation of fission rate reduces from -0.97% to -0.53%. Considering the resonance of zirconium in cladding can improve the accuracy of kinf by about 60 pcm. © 2018, Editorial Board of Atomic Energy Science and Technology. All right reserved.

Keyword :

Calculation results Cross section data Data library Fission rates Lattice codes Multiplication factor Reference values Relative deviations

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GB/T 7714 Wang, Dongyong , Wu, Hongchun , Li, Yunzhao et al. Improvement of Multi-group Cross Section Data Library of Bamboo-Lattice Code [J]. | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology , 2018 , 52 (8) : 1367-1373 .
MLA Wang, Dongyong et al. "Improvement of Multi-group Cross Section Data Library of Bamboo-Lattice Code" . | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology 52 . 8 (2018) : 1367-1373 .
APA Wang, Dongyong , Wu, Hongchun , Li, Yunzhao , He, Qingming . Improvement of Multi-group Cross Section Data Library of Bamboo-Lattice Code . | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology , 2018 , 52 (8) , 1367-1373 .
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Unstructured coarse mesh finite difference method to accelerate k-eigenvalue and fixed source neutron transport calculations EI SCIE Scopus
期刊论文 | 2018 , 120 , 367-377 | ANNALS OF NUCLEAR ENERGY
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Unstructured coarse mesh finite difference (CMFD) method was implemented to accelerate the k-eigenvalue and fixed source neutron transport calculations in the nodal S-N transport code DNTR, which was based on arbitrary triangular-z meshes. The same triangular-z meshes were adopted in the CMFD calculation to accommodate the arbitrary problem boundaries and simplify the coarse mesh generation and mapping processes. Two-level CMFD formulations were derived for the k-eigenvalue problems, and the few-group CMFD calculation was used to speed up the convergence of the multigroup CMFD calculation. For the fixed source problems in the neutronics analyses of subcritical reactors or transient neutron transport calculations, a special k(s), iteration method by introducing the source multiplication factor ks was employed in the S-N transport calculation to remove the dependence of the convergence rate on the subcriticality. Multigroup CMFD formulations were established to accelerate the convergence of the k(s), iteration procedure in the fixed source neutron transport calculation. The acceleration effects of CMFD on the k-eigenvalue and fixed source calculations were verified using four representative neutron transport problems, i.e. the small fast breeder reactor with Cartesian boundaries, the BN-600 fast reactor with hexagonal boundaries, the space nuclear power reactor with cylindrical boundaries, and the accelerator driven subcritical reactor with hexagonal boundaries. The numerical results showed that for the first three eigenvalue problems, the multigroup CMFD achieved a speedup of total computation time by a factor of 2-5, and the two-level CMFD reduced the computation time by a factor of 3-7. It was also shown that the multigroup CMFD was able to accelerate the k(s) iteration procedure in the fixed source problems by a factor of about 6. The implementation of the unstructured CMFD method turns the DNTR code into an efficient neutron transport solver with flexible geometry treatment capability for both k-eigenvalue and fixed source problems. (C) 2018 Elsevier Ltd. All rights reserved.

Keyword :

Acceleration Unstructured coarse mesh finite difference k-Eigenvalue neutron transport Fixed source neutron transport

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GB/T 7714 Zhou, Shengcheng , Wu, Hongchun , Cao, Liangzhi et al. Unstructured coarse mesh finite difference method to accelerate k-eigenvalue and fixed source neutron transport calculations [J]. | ANNALS OF NUCLEAR ENERGY , 2018 , 120 : 367-377 .
MLA Zhou, Shengcheng et al. "Unstructured coarse mesh finite difference method to accelerate k-eigenvalue and fixed source neutron transport calculations" . | ANNALS OF NUCLEAR ENERGY 120 (2018) : 367-377 .
APA Zhou, Shengcheng , Wu, Hongchun , Cao, Liangzhi , Zheng, Youqi . Unstructured coarse mesh finite difference method to accelerate k-eigenvalue and fixed source neutron transport calculations . | ANNALS OF NUCLEAR ENERGY , 2018 , 120 , 367-377 .
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Feasibility Study of Heterogeneous Resonance Integral Table of 238U EI Scopus CSCD PKU
期刊论文 | 2018 , 52 (7) , 1174-1180 | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology
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As the complexity and accuracy requirement increase of neutronics calculation of the next generation reactor, it is necessary to introduce the heterogeneous resonance integral table into the multi-group cross section library replacing the homogeneous resonance integral table. The accuracy of embedded self-shielding calculation using heterogeneous resonance integral tables of 238U prepared in different schemes was analyzed. The results show that the density of the moderator and the concentration of boron are crucial in the practical application of heterogeneous resonance integrals. Based on the analysis results, the current scheme of the 2D heterogeneous resonance integral table was improved by a 4D interpolation table to cover all possible operating conditions of PWR. Numerical results show that the improved heterogeneous resonance integral table enhances the accuracy of resonance calculation and improves the applicability of the multi-group library. © 2018, Science Press. All right reserved.

Keyword :

Cross-section library Feasibility studies Heterogeneous systems Multi-group Numerical results Operating condition Resonance integral Self-shielding calculation

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GB/T 7714 Zhang, Qian , Li, Song , Zhao, Qiang et al. Feasibility Study of Heterogeneous Resonance Integral Table of 238U [J]. | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology , 2018 , 52 (7) : 1174-1180 .
MLA Zhang, Qian et al. "Feasibility Study of Heterogeneous Resonance Integral Table of 238U" . | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology 52 . 7 (2018) : 1174-1180 .
APA Zhang, Qian , Li, Song , Zhao, Qiang , Wu, Hongchun , Zhuang, Kun . Feasibility Study of Heterogeneous Resonance Integral Table of 238U . | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology , 2018 , 52 (7) , 1174-1180 .
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Subgroup Resonance Calculation Method under Rim Effect in Fuel Rod EI Scopus CSCD PKU
期刊论文 | 2018 , 52 (8) , 1374-1380 | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology
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Abstract :

Pseudo-resonant-nuclide subgroup method (PRNSM) is capable to treat spatially self-shielding effect in fuel rod. However, it requires the number densities of resonance nuclides to be constant throughout the whole fuel region, so the spatial distributions of resonance nuclides' densities caused by rim effect cannot be taken into account. To solve this problem, a new method was proposed on the base of improved pseudo-resonant-isotope-theory. Numerical results show that this method solves the limit of PRNSM on rim effect and has higher accuracy than Bondarenko iteration method and resonance interference factor method. © 2018, Editorial Board of Atomic Energy Science and Technology. All right reserved.

Keyword :

Fuel rods Interference factor Iteration method Number density Numerical results Self shielding effect Self-shielding Subgroup method

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GB/T 7714 Zhang, Qian , Li, Song , Zhao, Qiang et al. Subgroup Resonance Calculation Method under Rim Effect in Fuel Rod [J]. | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology , 2018 , 52 (8) : 1374-1380 .
MLA Zhang, Qian et al. "Subgroup Resonance Calculation Method under Rim Effect in Fuel Rod" . | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology 52 . 8 (2018) : 1374-1380 .
APA Zhang, Qian , Li, Song , Zhao, Qiang , Song, Peitao , Cao, Liangzhi , Wu, Hongchun . Subgroup Resonance Calculation Method under Rim Effect in Fuel Rod . | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology , 2018 , 52 (8) , 1374-1380 .
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Development and Validation of Nuclear Data Processing Code NECP-Atlas EI Scopus CSCD PKU
期刊论文 | 2018 , 52 (7) , 1160-1165 | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology
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A home-developed nuclear data processing code named as NECP-Atlas was described in this paper, which included different code modules for different data procession functions. This code can generate WIMS-D or WIMS-E format multigroup library from evaluated nuclear data file, through reconstruction of resolved resonance cross section and linearization of cross sections, Doppler broadening, calculation of unresolved resonance cross sections, calculation of thermal neutron scattering cross sections, calculation of multigroup cross section, and so on. The NECP-Atlas code was tested against WIMSD library update project (WLUP) benchmark and international criticality safety benchmark evaluation project (ICSBEP). The numerical results show that the NECP-Atlas code can get results comparable with NJOY-2016 code. © 2018, Science Press. All right reserved.

Keyword :

Benchmark evaluation Criticality safety Evaluated nuclear data file Multigroup cross sections Numerical results Resonance cross-section Thermal neutron scattering Thermal neutron scattering cross section

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GB/T 7714 Zu, Tiejun , Xu, Jialong , Wu, Hongchun et al. Development and Validation of Nuclear Data Processing Code NECP-Atlas [J]. | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology , 2018 , 52 (7) : 1160-1165 .
MLA Zu, Tiejun et al. "Development and Validation of Nuclear Data Processing Code NECP-Atlas" . | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology 52 . 7 (2018) : 1160-1165 .
APA Zu, Tiejun , Xu, Jialong , Wu, Hongchun , Cao, Liangzhi . Development and Validation of Nuclear Data Processing Code NECP-Atlas . | Yuanzineng Kexue Jishu/Atomic Energy Science and Technology , 2018 , 52 (7) , 1160-1165 .
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